1. Field of the Invention
This invention relates to an operating method for a nuclear power plant, and more particularly to an operating method for a nuclear power plant capable of maintaining continuous operation of the nuclear reactor by scraming the previously selected control rods (hereinafter referred to as the selected control rods) when an in-house or outside power failure or decrease in power occurs which results in a reduction in demand of a boiling water reactor (hereinafter referred to as BWR) type nuclear power plant.
2. Description of the Prior Art
The technical term "scram" means rapid-insertion of the control rods.
In general, a continuous isolated grid operation (in-house separate operation) is maintained during an electric grid system failure, by an operating method capable of maintaining continuous operation of the nuclear reactor without an interruption thereof. This can be achieved, upon occurrence of the failure, by tripping primary loop recirculation system pumps, by scraming the selected control rods so as to reduce the nuclear reactor power level, and by rapidly opening the turbine bypass valves so as to bypass steam generated from the nuclear reactor into a condenser. Then the nuclear reactor is operated from about 20% to 30% of its rated power level.
However, the selected control rods subjected to the scram process are required to be set by an operator at every burnup point in accordance with the result of nuclear-thermal hydraulic analysis previously made at every burnup point. Moreover, in case of power recovery of the nuclear power plant after repair of the failure, the selected control rods which were subjected to the scram process should be withdrawn only after the nuclear reactor power has further been lowered in order to conform with operation procedures for fuel preconditioning, because the nuclear reactor power distributions are inevitably distorted resulting in adverse influence on the fuels, and at the same time such cumbersome and complicated control rod sequences deteriorate the availability factor of the nuclear power plant.
In the conventional selected control rod inserting method, there are also such disadvantages that the inserting method is implemented only upon the occurrence of an outside power failure affecting the nuclear power plant, but not in the case of an in-house power failure, (i.e., upon occurrence of a failure such as an accidental stoppage of the nuclear reactor feedwater pump, the nuclear power plant becomes incapable of maintaining continuous operation of the generator along with an abrupt decrease in the nuclear reactor power level).
Furthermore, in a status under which the selected control rod insertion has determined the natural circulation of the coolant and a lowered power level, there are also other well-known disadvantages such that the axial power distribution of the reactor core is downwardly distorted due to the following three causes:
(i) In the case of a lower flow rate within the core, such as natural circulation of the coolant. Voids generated in the upper part of the core are not sufficiently swept away, so that neutron slow-down effects are insufficiently obtained, consequently the power developed in the lower part of the core becomes correlatively greater.
(ii) The thermal source of a feedwater heater is supplied by steam extracted from a turbine; however, when such steam is bypassed into a condenser by abrupt opening of the turbine bypass valve, the thermal source of the feedwater heater is interrupted, so that water with relatively higher subcooling is fed into the core. This causes the axial boiling start-up point within the core to be raised. As a result, neutron slow-down effects in the lower part of the core are enhanced, thus the power in the lower part of the core is increased.
(iii) As shown in FIGS. 1 and 2, the higher the void quantity in the core, the greater the control rod reactivity, so that when the selected control rods are fully inserted, the negative reactivity effects caused by the selected control rods become greater in the upper part of the core and smaller in the lower part of the core. Therefore, when the selected control rods are subjected to a scram process to a fully inserted position, the reactor power distribution becomes downwardly distorted.
As described above, there are several interactively related causes, so that when the selected control rods are subjected to a scram process, the axial power distribution tends to be excessively downwardly distorted.
When the power peaking in the lower part of the core becomes greater, this adversely affects the core channel stability, which will be described in greater detail hereinafter.
When analyzing the stability of such a large-scale nonlinear type system as a BWR type nuclear power plant, the stability of the constituent elements or subsystems should first be examined. Then the stability of the whole system consisting of combinations of such elements and subsystems should be examined.
For example, the thermal hydraulic stability of each individual channel's (flow passage) within the core may be examined to confirm the inherent stability (i.e. the channel stability) thereof.
Next, these channel stabilities are hydraulically combined, further combined with the nuclear characteristics and thermal transfer characteristics within the core, and then the stability of the core (i.e. "core stability") will be examined.
The channel stability will be explained with respect to a BWR type nuclear reactor which is provided with several hundreds of fuel assemblies installed within the core thereof, such that the respective fuel channels are arranged to form parallel channels. With such a design, the influence of the flow rate oscillation from even a single channel is absorbed by numerous channels surrounding the flow oscillation generating channel, so that no change in pressure appears at the inlet or the outlet of the core.
In the above-described heated two-phase flow channel, it may be recognized that even when the quantity of heat is constant, flow oscillation thereof can be produced.
The stability of such a two-phase flow channel has so far been studied in various aspects, and the instabilities of various kinds have been recognized, and also systematically classified.
In accordance with such classifications, one of the most common types of wave oscillations which affects channel stability and which will now be discussed is a so-called density wave oscillation. The mechanism of such oscillation is, in short, such that it is derived from the transfer delay and the feedback effects of such variables present within the channel as flow rate, density (void fraction) and pressure loss. One feature of such a mechanism is that the period of oscillation has a close relationship to the time in which the density waves within the flow (or propagation waves of void fraction disturbances) pass through the channels. Such oscillation was previously known as flow-rate void-feedback instability or time-delay instability, however, on the basis of the aforementioned features, is presently known as density wave oscillation.
FIG. 3 shows a prior art BWR fuel channel provided with an inlet, a single phase flow portion, a subcooled boiling portion, a bulk boiling portion and an outlet. In the subcooled boiling portion, although enthalpy of water has not reached saturation enthalpy thereof, there exist steam bubbles, while in the bulk boiling portion, water has reached the saturation enthalpy thereof. Hereinafter, the mechanism of the state in which the oscillation occurs will be described in more detail.
Although for simplicity's sake, the subcooled boiling portion is omitted, no problem exists because the discussion is made in regard to a qualitative understanding.
Now, assume that the inlet flow rate of the channel is in oscillation, and that this oscillation produces a propagation along the flow of enthalpy disturbance with respect to the single phase flow portion. The boiling boundary (hereinafter refer to as BB) at which the temperature of the water reaches saturation oscillates because of the enthalpy disturbance. Since the flow rate and the length of the single phase flow portion oscillate, pressure loss within the single phase flow portion also oscillates. The oscillation of BB, that is, the oscillation of the void fraction or quality thereat, propagates along the flow, and at the same time, causes the flow velocity within the two-phase flow portion to produce disturbance. The void fraction and flow velocity disturbances, in conjunction with the length of oscillation of the two-phase flow portion, interactively produce the pressure loss disturbance within the two-phase flow portion.
Here, the entire pressure loss of the channel is given externally as a boundary condition, which is, in this case, a constant determined by several hundreds of other channels. Thus, the pressure loss disturbance within the two-phase flow portion effects the single-phase flow portion by a pressure loss change identical thereto in magnitude and reverse thereto in polarity. This enhances the firstly assumed hypothetical oscillation in the inlet flow velocity (in case of instability), or attenuates the same (in case of stability).
More detailed studies as to the case of a critical oscillation will be described hereinafter. In this case, a pressure loss change within the single-phase flow portion becomes identical in magnitude and reverse in polarity to that within the two-phase flow portion. Thus, with respect to the operating conditions of the BWR, the pressure loss change of the single phase flow portion is substantially in-phase with respect to the inlet flow rate change, while on the other hand, the pressure loss change of the two-phase flow portion is substantially in-phase with respect to the outlet flow rate change. Therefore, the flow rate, in this case, has a considerable amount of delay between the inlet phase and the outlet phase. Such a phase delay does not appear in a non-compressible single phase flow, and it is caused by a greater density change along the flow within the boiling channel. Namely, water that entered the channel with a greater inlet flow rate should pass through a longer distance before it reaches a saturated temperature because of the greater velocity thereof. Consequently, BB shifts toward the downstream. In the boiling portion, this is propagated as the negative void fraction disturbance, thereby causing the density flow rate to be propagated as a positive disturbance due to the density difference between water and steam.
As a result, this functions to increase the pressure loss of the boiling portion, to decrease the pressure loss of the single phase flow portion, and to decrease the inlet flow rate. This explains why half the period of oscillation becomes substantially equal to the time during which the fluid passes through the channel.
Under such conditions, should the pressure loss of the two-phase flow portion be increased, the pressure loss change of the two-phase flow portion is increased; and thereby increases the instability of the channel. On the other hand, a decrease in the size of the opening in the channel inlet orifice lowers the gain of the inlet flow rate change derived from the pressure loss change of the single-phase flow portion which is caused by the pressure loss change of the boiling portion, thereby causing stability to be enhanced. These tendencies have experimentally been confirmed.
When power peaking in the lower part of the core becomes greater, the void factor is increased, the pressure loss of the two-phase flow portion becomes greater, and the instability of the channel is increased.
Next, a detailed explanation of core stability will be described.
After the channel stability is assured, the core stability of a core having several hundreds of collective channels may be examined. In this case, it may be considered that the thermal hydraulic characteristics of the channels and the nuclear reactor core characteristics derived from the void reactivity coefficient are combined, thereby causing the core to become unstable. Here, the whole core is simulated by grouping the numerous fuel channels into several groups whose thermal-hydraulic characteristics are similar.
The thermal-hydraulic dynamic characteristics as to the respective fuel channels are derived from mathematical models of the previously described channel stability. Here, the channel flow rate and the void factor are produced with respect to the channel pressure loss and the input of thermal flow bundle. Now, the sum of channel flow rates becomes the core flow rate, and the dynamic characteristic models with respect to the primary loop recirculation system are required so that the pressure change of the core inlet plenum may be obtained. Finally, the void fraction of the respective channels are multiplied by the void reactivity coefficient, and summed up, then the reactivity change of the whole core is obtained, which, in turn, becomes the input of the core thermal characteristic and constitutes a feedback loop.
Therefore, in this case, the stability of the core is similar to that of the conventional feedback system, and readily appreciable.
Similar to the case of channel stability, when the axial power peaking in the lower part of the core becomes greater, the void fraction increases, and the feedback gain of the void reactivity is enhanced, whereby the instability of the core is increased.
As described above, when the axial power peaking in the lower part of the core becomes greater, the void quantity within the channel of the core increases, and the pressure loss and the void quantity of the two-phase flow is then increased. As a result, when minute disturbances occur within the channel flow rate, the flow rate disturbance delay caused by the void becomes greater, then the oscillation tends to readily continue. Also with regard to reactivity, the void reactivity is increased and a tendency towards instability emerges.